ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 24

Document Title:

Lawrence Livermore National Laboratory – 

Occupational External Dose

Document Number: ORAUT-TKBS-0035-6 

Revision: 01 

Effective Date: 04/26/2007

Type of Document TBD

Supersedes: Revision 00

Subject Expert(s): Bill R. Thomas and Paul A. Szalinski

Site Expert(s): N/A

Approval: Signature on File Approval Date: 04/24/2007

Paul A. Szalinski, Document Owner

Concurrence: Signature on File Concurrence Date: 04/24/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 04/24/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 04/24/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 04/26/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 2 of 24

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

10/07/2005 00 New technical basis document for the Lawrence Livermore National 

Laboratory – Occupational External Dose. First approved issue. 

Initiated by Jay J. Maisler.

04/26/2007 01 Approved revision to respond to comments by ORAU and Task 5. 

This revision specifically does not respond to Worker Outreach 

comments received; which will be addressed in a future revision. 

Constitutes a total rewrite of the document. Incorporates formal 

internal and NIOSH review comments. Adds Attributions and 

Annotations section. This revision results in no change to the 

assigned dose and no PER is required. Training required: As 

determined by the Task Manager. Initiated by Paul A. Szalinski.
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 3 of 24

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

5

6.1 Introduction ..............................................................................................................................6

6.1.1 Purpose.....................................................................................................................7

6.1.2 Scope ........................................................................................................................7

6.1.3 Basis of Comparison..................................................................................................7

6.2 Dose Reconstruction Parameters.............................................................................................8

6.2.1 Historical Administrative Practices.............................................................................8

6.2.2 Dosimetry Technology .............................................................................................10

6.2.2.1 Beta/Gamma Dosimeters ....................................................................... 10

6.2.2.2 Neutron Dosimeters ............................................................................... 10

6.2.3 Calibration ...............................................................................................................11

6.2.3.1 Beta/Photon Dosimeters......................................................................... 11

6.2.3.2 Neutron Dosimeters ............................................................................... 11

6.2.4 Workplace Radiation Fields .....................................................................................11

6.2.5 Recorded Dose Practices ........................................................................................11

6.2.6 Interpretation of Reported Data ...............................................................................11

6.3 Adjustments to Recorded Dose..............................................................................................14

6.4 Missed and Unmonitored Dose ..............................................................................................15

6.5 Bias and Uncertainty..............................................................................................................15

6.6 Dose Reconstruction..............................................................................................................16

6.7 Organ Dose............................................................................................................................17

6.8 Attributions and Annotations ..................................................................................................17

References ........................................................................................................................................20

Glossary ............................................................................................................................................22
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 4 of 24

LIST OF TABLES

TABLE TITLE

6-1 LLNL dosimeter type, period of use, exchange frequency, MDL, and 

potential annual missed dose...................................................................................................

PAGE

9

6-2 Selection of beta and photon radiation energies and percentages ......................................... 12

6-3 Selection of neutron radiation energies and percentages....................................................... 13

6-4 Recorded dose practices........................................................................................................ 13

6-5 Interpretation of reported data................................................................................................ 14

6-6 Adjustments to recorded dose................................................................................................ 15

6-7 Bias and uncertainty...............................................................................................................16
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 5 of 24

ACRONYMS AND ABBREVIATIONS

CFR Code of Federal Regulations

CR-39 Columbia Resin – 39 or as polyallyl diglycol carbonate (PADC or CR-39), a 

polycarbonate material used for neutron dosimetry

DCF dose conversion factor

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act

GSD Geometric standard deviation

Hp(d) personal dose equivalent at tissue depth d (d = 10 mm or 0.07 mm)

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

keV kilovolt-electron

LBNL Lawrence Berkeley National Laboratory

LANL Los Alamos National Laboratory

LLNL Lawrence Livermore National Laboratory

MDL minimum detection level

MED Manhattan Engineer District

MeV megaelectron-volt

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-Office of Compensation Analysis and Support Claims Tracking System

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

ORAU Oak Ridge Associated Universities

POC probability of causation

RFETS Rocky Flats Environmental Technology Site

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

WB whole body
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 6 of 24

6.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c)...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 7 of 24

6.1.1

This Site Profile documents historical practices at the Lawrence Livermore National Laboratory (LLNL) 

and provides information for the evaluation of external dosimetry data for monitored workers; it can 

serve as a supplement to, or substitute for, individual monitoring data.

Purpose

This document provides a uniform and consistent approach to assessing occupational external dose 

at LLNL for dose reconstructions in relation to the EEOICPA. It provides guidance to dose 

reconstructors on input parameters that are specific to LLNL employees, as well as the approach for 

employees with either missing or no monitoring information.

LLNL workers handled a variety of radionuclides as part of their routine work. The key elements in 

the source term were plutonium and tritium although others were used at various times and in various 

forms. For the purposes of dose reconstruction, it can be assumed that internal source terms were 

introduced at LLNL’s inception on September 2, 1952.

6.1.2

Workers at LLNL were exposed to radiation from a variety of radioactive materials and radiationproducing machines. In addition, many LLNL workers worked at the Nevada Test Site or were 

involved with other weapons tests where they could have received radiation exposures. Personnel 

dosimeter records are generally available for all periods at LLNL for workers who had any potential for 

occupational radiation exposure. The operations and radiation safety staff routinely reviewed 

dosimeter results for compliance with radiation control limits and investigated doses approaching 

annual or quarterly dose limits. 

Scope

Radiation dosimetry practices were initially based on experience gained during several decades of 

radium and X-ray medical diagnostic and therapy applications. 

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 6.8.

6.1.3

Various radiation dose concepts and quantities have been in use to measure and record occupational 

dose since the initiation of the MED in the early 1940s. The basis of comparison for reconstruction of 

dose is the personal dose equivalent, Hp(d), where d identifies the depth in millimeters and represents 

the point of reference for dose in tissue. Hp(d) is an operational quantity used to determine the 

worker’s actual dose equivalent from a dosimeter reading. International Commission on Radiological 

Protection (ICRP) Publication 74 (ICRP 1996) and International Commission on Radiation Units and 

Measurements (ICRU) Report 57 (ICRU 1998) both define it as a practical method for calibration of 

instruments and dosimeters to dose equivalent. Weakly penetrating radiation is significant to shallow 

dose equivalent, which is defined at d equal to 0.07 mm and noted as Hp(0.07). Penetrating radiation 

is significant to deep dose equivalent, which is defined at d equal to 10 mm and noted as Hp(10). 

Hp(0.07) and Hp(10) are the radiation quantities used in the DOE Laboratory Accreditation Program 

(DOELAP) used to accredit DOE personnel dosimetry systems since the 1980s (DOE 1986). The 

National Voluntary Laboratory Accreditation Program (NVLAP), which is the U.S. Nuclear Regulatory 

Commission equivalent to DOELAP, uses the same operational quantities. This TBD uses Hp(10)

and Hp(0.07) as deep and shallow dose equivalents, respectively.

Basis of Comparison
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 8 of 24

6.2 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (X-ray and gamma ray), and neutron radiation type, energy, and 

geometry of exposure in the workplace, and the characteristics of the LLNL dosimeter responses are 

important to the assessment of bias and uncertainty of the original recorded dose in relation to the 

radiation quantity Hp(d). Dose reconstructors can compare earlier dosimetry systems to current 

systems to evaluate their performance based on the premise that current systems have more 

stringent criteria, as indicated in the DOELAP and NVLAP programs.

Accuracy and precision of the recorded individual worker doses depend on (Fix et al. 1997):

• Administrative practices

•

that facilities adopt to calculate and record personnel dose based on 

technical, administrative, and statutory compliance considerations. 

Dosimetry technology

•

, which includes the physical capabilities of the dosimetry system, such 

as the response to different types and energies of radiation, in particular in mixed radiation 

fields. 

Calibration

•

of the respective monitoring systems and similarity of the methods of calibration to 

sources of exposure in the workplace. 

Workplace radiation fields

An evaluation of the original recorded doses, as available, combined with detailed examinations of 

workplace radiation fields and dosimeter responses to those fields is the recommended option to 

provide the best estimate of Hp(d) for individual workers.

, which can include mixed types of radiation, variations in exposure 

geometries, and environmental conditions. 

6.2.1

LLNL used personnel dosimeters to measure and record doses from external radiation to designated 

workers throughout the history of its operations. These dosimeters include one or more of the 

following:

Historical Administrative Practices

• Personnel whole-body (WB) beta/photon dosimeters 

• Pocket ionization chamber dosimeters (i.e. electrometers, etc.)

• Personnel extremity dosimeters

• Personnel neutron dosimeters

LLNL began operations in 1952 using dosimeter and processing technical support provided by the 

Lawrence Berkeley National Laboratory (LBNL), which was then known as the University of California 

Radiation Laboratory. LBNL had implemented its dosimetry methods based on the personnel beta/ 

photon dosimeter design developed at the Metallurgical Laboratory at the University of Chicago 

(Pardue, Goldstein, and Wollan 1944). From 1952 to 1955, LBNL provided LLNL with photon/electron 

film dosimeters and nuclear track emulsion, type A (NTA) film; technical details regarding the 

capabilities of the external dosimetry system are provided in the LBNL site profile (ORAUT 2006a). 

By 1956, LLNL had its own fully functional personnel dosimetry program with in-house processing. 

Early exposure records also provided “electrometer” results (designated by “E” in selected claimant 

files), which supplemented the official results measured by film. For purposes of dose reconstruction, 

the “electrometer” or “E” results should be used in a qualitative manner because no data was found 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 9 of 24

on the calibration or energy response of these devices; the film or TLD results should be used to 

estimate the actual exposure.

Table 6-1 summarizes the personnel dosimeters used at LLNL over the years, along with their periods 

of use, exchange frequencies, minimum detection levels (MDLs), and estimated annual missed 

doses. 

Table 6-1. LLNL dosimeter type, period of use, exchange frequency, MDL, and potential annual

missed dose [1]. 

Period of use Dosimeter

MDLa 

(rem)

Exchange

frequencyb

Annual 

missed 

dosec (rem)

1952–1969d Photon/electron–DuPont 558 and 519 film 0.015 Weekly (n = 50) 0.375

0.020 Biweekly (n = 25) 0.25

0.030 Monthly (n = 12) 0.18

Neutron–Kodak NTA film 0.050 Weekly (n = 50) 1.25

0.050 Biweekly (n = 25) 0.625

0.050 Monthly (n = 12) 0.30

1969–1985 Photon/electron/neutron - Harshaw TLD 

(TLD-100, TLD-200, TLD-600, and TLD-700)

0.010 Monthly (n = 12) 0.06

0.020 Quarterly (n = 4) 0.04

1985–present Photon/electron–Panasonic 810AS and 802AS 

TLD

0.010 Monthly (n = 12) 0.06

0.015 Quarterly (n = 4) 0.03

0.025 Semiannual (n = 2) 0.025

Neutron–CR-39 0.010 Monthly (n = 12) 0.06

0.010 Quarterly (n = 4) 0.02

a. Estimated MDLs for each dosimeter technology in the workplace. Dose values were recorded at levels less than the 

MDL. 

b. Exchange frequencies were dependent on work assignment. If the exchange frequency is not evident based on trends 

in an individual’s personnel records, assume a monthly exchange frequency.

c. Annual missed dose calculated using the MDL/2 method from NIOSH (2006).

d. From 1952 to 1955, LBNL processed film dosimeters. Dosimetry records in LLNL database prior to 1952 were 

processed by LBNL.

Dose reconstruction parameters concerning LLNL administrative practices significant to dose 

reconstruction involve policies to:

• Assign dosimeters to workers. 

• Exchange dosimeters. 

• Estimate dose for missing or damaged dosimeters. 

• Replace destroyed or missing records. 

• Evaluate and record dose for incidents. 

• Obtain and record occupational dose to workers for other employer exposure. 

LLNL policies were in place for all these parameters. From its inception, LLNL had policies to monitor 

individuals with any significant potential for radiation exposure. The current practice of monitoring all 

workers entering the site, regardless of exposure potential, has been in effect since May 1958, when 

film badges became part of the security badge (Nolan 1958). Dosimeter exchange frequencies varied 

over the years depending in part on the dosimeter type in use at the time and in part on the exposure 

potential of the individual being monitored; individuals with low exposure potential tended to have less 

frequent exchange frequencies than those with high exposure potential. Doses below the detection 

limit were either recorded as zero or not recorded at all; a review of representative claimant files 

indicates some inconsistencies. It was apparent that some individuals did not return their dosimeter 

for processing; the record indicates the term “missing.” 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 10 of 24

6.2.2

The LLNL dosimetry methods evolved during the years as improved technology was developed and 

the complex radiation fields encountered in the workplace were better understood. The adequacy of 

the respective dosimetry methods to measure radiation dose accurately as discussed in later sections 

depends on radiation type, energy, exposure geometry, etc. The exchange frequency of the 

dosimeters was gradually lengthened and corresponded generally to downward reductions in the 

radiation protection guidelines (Morgan 1961; Taylor 1971). The dosimeter designs accommodated 

the numerous radiation field types that workers might encounter throughout the LLNL complex. 

Dosimetry Technology

6.2.2.1 Beta/Gamma Dosimeters

From 1952 through 1955, LBNL supplied and processed beta/gamma dosimeters. These film 

dosimeters were similar in design to those developed at the Metallurgical Laboratory at the University 

of Chicago (Pardue, Goldstein, and Wollan 1944). LLNL used a similar design when it began its inhouse dosimetry program in 1956. These film dosimeters, which were in use until 1969, used DuPont 

558 and 519 film.

In 1969, thermoluminescent dosimeters (TLDs) replaced film dosimeters. LLNL constructed the TLDs 

used from 1969 through 1985, which contained Harshaw TLD-100, TLD-200, TLD-600, and TLD-700

elements. The use of Panasonic 802 and 810 dosimeters began in 1985 and has continued to the 

present. 

6.2.2.2 Neutron Dosimeters 

From 1952 through 1955, LBNL supplied and processed neutron dosimeters, which used Kodak NTA 

film. LLNL-provided dosimeters used the same film from 1956 through 1969. At some atomic 

weapons employer facilities or DOE facilities, neutron dosimeters were not read out unless a 

minimum gamma exposure threshold was reached. There is no evidence that this was practiced at 

LLNL. In Personnel Monitoring Procedure for UCRL, Livermore (LRL ca. 1954) it is stated that: “A film 

badge containing fast neutron sensitive film shall also be provided for persons who will be working 

within areas where significant neutron exposures are possible.” No definition of "significant" was 

provided; at this time there are no clear criteria for how neutron dosimeters were assigned. If 

available, coworker data should be used. It is clear that all unmonitored workers did not have the 

potential for exposure to neutrons based on buildings worked in and job classifications, but the criteria 

for monitoring has not been quantified. Research into dosimetry practices at LBNL to develop the 

TBDs may provide additional information on dosimetry practices. This document will be revised if 

additional evidence is found.

As far as badge exchange and processing, the document states that: “For those who work in radiation 

areas, the film badges are exchanged once a week and fresh film is installed. The old film is 

developed and the radiation dosage is recorded.” Nowhere in this document is it mentioned that not 

all neutron film badges were processed. However, only values above the limit of detection were 

reported.

In 1969, TLDs replaced film dosimeters. LLNL constructed the TLDs used from 1969 through 1985, 

which contained Harshaw TLD-100, TLD-200, TLD-600, and TLD-700 elements. From 1969 through 

1975, the neutron dosimeter used a non-Albedo-type design. Badge modifications in 1975 used the 

Albedo design whereby cadmium shielded the TLD-600 and TLD-700 components on all sides except 

the side facing the wearer’s body. This design effectively shielded incident thermal neutrons from the 

TLD-600 and TLD-700 elements and detected only thermal neutrons reflected from the wearer’s body. 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 11 of 24

The change in design did not modify the detection limit. The change provided a more precise result 

compared to the over-estimate which was observed prior to 1975. The design, which minimized the 

over-response of the earlier design, was in use until 1985 when LLNL adopted Panasonic TLDs. Also 

in 1985, Columbia Resin-39 (CR-39) neutron dosimeters were introduced and used with the 

Panasonic TLDs. This system continues in use at present.

6.2.3

Potential error in recorded dose is dependent on the dosimeter response characteristics to each 

radiation type, energy, and geometry; the methodology used to calibrate the dosimetry system; and 

the extent of similarity between the radiation fields used for calibration and the field present in the 

workplace. The potential error is much greater for dosimeters with significant variations in response, 

such as film dosimeters for low-energy photon radiation and the NTA film and TLDs for neutron 

radiation.

Calibration

6.2.3.1 Beta/Photon Dosimeters

Dosimeters were calibrated using 226Ra, 137Cs, or 60Co in air (i.e., no phantom) until the 1986 adoption 

of DOELAP procedures requiring calibration with phantoms. Since 1986, photon calibrations have 

been to 137Cs with a phantom. Beta calibrations were routinely to 90Sr.

6.2.3.2 Neutron Dosimeters

Neutron dosimeters were calibrated using PuBe sources prior to 1970. Since 1970, unmoderated and 

moderated 252Cf has been used.

6.2.4

The radiation fields at LLNL are highly variable. They include radiation from a variety of radiationproducing machines such as electron accelerators, X-ray machines, cyclotrons, neutron generators, 

and a research nuclear reactor. In addition, many different radioactive materials have been used at 

LLNL. Tables 6-2 and 6-3 list many of the radiation sources that might have been encountered at 

LLNL over the years. All beta particle energies are greater than 15 keV. In general, beta and/or 

charged particle exposure around the accelerators is limited to shallow dose. Due to the manner in

which accelerated charged particles were contained, individuals would not have been exposed to 

energetic charged particles. The neutron energy selection of 0.1 – 2.0 MeV indicated in Table 6-3 

was chosen because neutron sources were shielded and moderated to this energy range; the energy 

range of 0.1 – 2.0 MeV has the highest quality factor, as provided in 10 C. F. R. § 835.2, which is 

favorable to claimants. 

Workplace Radiation Fields

6.2.5

Table 6-4 summarizes the terminology used to describe dosimeter measured and compliance dose 

quantities. 

Recorded Dose Practices

6.2.6

Dose reconstructors can use the information in Table 6-5 to assist in the interpretation of LLNL 

external dosimetry summary reports. Because much of the data prior to 1958 was not adequately 

defined to directly report shallow and neutron doses, the dose reconstructor should use the guidance 

found within External Dose Reconstruction (ORAUT 2006b), Interpretation of Dosimetry Data for 

Interpretation of Reported Data
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 12 of 24

Table 6-2. Selection of beta and photon radiation energies and percentages [2]. 

Buildings–

old 

numbersa 

Buildings–

new 

numbersa Description

Radiation 

type

Energy 

selection (keV) Percentage

101, 102, 

106, 117, 

118, 147, 

176, 192

222, 221, 223, 

224, 234, 232, 

233, 167, 168, 

169

Chemistry: radioactive materials 

including Co-60, fission products, 

enriched uranium, depleted 

uranium, natural uranium, U-233, 

Cm-244, Pu-239, Am-241, others

Beta >15 100

Photon 30–250 

>250

50

50 

153, 154, 

157, 173, 

180, 194

210, 212; 171, 

173-177; 241, 

243; 435, 442, 

443; 194

Physics: accelerators, activation 

products, H-3, others

Beta >15 100

Photon 30–250 

>250

25

75

103, 114, 

125, 127, 

174, 175

215, 321, 419, 

514, 243, 253

Lab Services: radioactive 

materials

Beta >15 100

Photon 30–250 

>250

75

25

110 261 Criticality Test Facility Beta >15 100

Photon 30–250 

>250

50

50

115 327 Radiography Beta >15 100

Photon 30–250 

250

25

75

121 412 Hot cells: high beta waste, Sr-90 Beta >15 100

Photon 30–250 

>250

75

25

170 131 Weapons Beta >15 100

Photon <30

30–250 

50

50

171 332 Metallurgical Chemistry: AKA 

Plutonium Facility

Beta >15 100

Photon <30

30–250 

50

50

172 331 Gaseous Chemistry: AKA Tritium 

Facility

Beta >15 100

Photon <30 (tritium brems) 100

182 162, 165, 166 Lab Services: 55 Ci Co-60 (1958) Beta >15 100

Photon 30–250 

>250

25

75

190 251 Chemistry Heavy Elements 

Facility: Cm-244, Am-241, U-233, 

Pu-239, others

Beta >15 100

Photon <30

30–250 

>250

25

50

25

193 281 Reactor Beta >15 100

Photon 30–250 

>250

25

75

Site 300 Site 300 Explosives Testing: linear 

accelerators, depleted uranium, H3, radiography

Beta >15 100

Photon 30–250 

>250

25

75

a. LLNL (2005). 

Assignment of Shallow Dose (ORAUT 2005a), and Assignment of Missed Neutron Doses Based on 

Dosimeter Records (ORAUT 2005b). A review of the claimant files indicated an inconsistency in the 

manner in which shallow dose was reported. In some cases, the reported shallow dose includes the 

shallow and penetrating dose. In other cases, the shallow dose excludes the penetrating dose. For 

purposes of dose reconstruction, it is assumed that if the reported shallow dose exceeds the 

penetrating dose, the shallow dose is the sum of the shallow and penetrating doses. If the reported 

shallow dose is less than the penetrating dose, the shallow dose should be derived as the sum of the 

shallow dose and the penetrating dose.
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 13 of 24

Table 6-3. Selection of neutron radiation energies and percentages [3]. 

Buildings– 

old numbersa Buildings–new numbersa Descriptionb Radiation type

Energy 

selection 

(MeV) Percentage

101, 102, 106, 

117, 118, 147, 

176, 192

222, 221, 223, 

224, 234, 232, 

233, 167, 168, 

169

Chemistry: radioactive materials including 

Co-60, fission products, enriched uranium, 

depleted uranium, natural uranium, U-233, 

Cm-244, Pu-239, Am-241, others

Neutron 0.1–2.0 100

153, 154, 157, 

173, 180, 194

210, 212; 171, 

173-177; 241, 

243; 435, 442, 

443; 194

Physics: accelerators, activation products, 

H-3, others

Neutron 0.1–2.0 100

103, 114, 125, 

127, 174, 175

215, 321, 419, 

514, 243, 253

Lab Services: radioactive materials Neutron 0.1–2.0 100

110 261 Criticality Test Facility Neutron 0.1–2.0 100

115 327 Radiography Neutron 0.1–2.0 100

121 412 Hot cells: high beta waste, Sr-90 Neutron 0.1–2.0 100

170 131 Weapons Neutron 0.1–2.0 100

171 332 Metallurgical Chemistry: AKA Plutonium 

Facility

Neutron 0.1–2.0 100

172 331 Gaseous Chemistry: AKA Tritium Facility Neutron 0.1–2.0 100

182 162, 165, 166 Lab Services: 55 Ci Co-60 (1958) Neutron 0.1–2.0 100

190 251 Chemistry Heavy Elements Facility: Cm244, Am-241, U-233, Pu-239, othersNeutron 0.1–2.0 100

193 281 Reactor Neutron 0.1–2.0 100

Site 300 Site 300 Explosives Testing: linear accelerators, 

depleted uranium, H-3, radiography

Neutron 0.1–2.0 100

a. Trost (2005). 

b. The facilities listed in this table represent the buildings where there was a potential for exposure to neutron radiation.

Table 6-4. Recorded dose practices[4]. 

Year

Dosimeter measured 

quantities Compliance dose quantities

Photon/electron film dosimeter + NTA neutron dosimeter

1952–1963 Beta (B)

Gamma (G)

Neutron (N)

WB = gamma + neutron + betaa

H = hand extremity dose (HN = hands) 

1963–1969 Beta (B)

Gamma (G)

Neutron (N)

WB = gamma + neutron

S = shallow (SK = skin) = gamma + neutron + beta

H = hand extremity dose (HN = hands) 

Photon/electron/neutron–Harshaw TLD

1969–1985 Beta (B)

Gamma (G)

Neutron (N)

WB = gamma + neutron

S = shallow (SK = skin) = gamma + neutron + beta

H = hand extremity dose (HN = hands) 

Photon/electron/neutron–Panasonic TLD + CR-39 neutron dosimeter

1985–present B/P/N Skin = NPEN + WB

WB = PEN + SN + FN

1995–2003 Shallow (SH or SK)

Deep photon (PH DP)

Deep neutron (NU DP)

Skin = beta + photon + neutron (B/P/N)

WB = photon deep + neutron deep

a. Prior to 1958, beta exposures were recorded as a component of the whole-body dose. 

The NTA dosimeter (1952-1969) exhibited a lower neutron energy threshold of approximately 500 keV

and consequently underestimated the neutron exposure (Griffith et al. 1979; NIOSH 2006; ORAUT 

2006c). The photon dose was measured adequately and all LLNL neutron dose was accompanied by 

a significant photon dose. For neutron dose received prior to 1969, the dose should be adjusted by 

using a neutron to photon ratio. The ratio varied by operation and task. There was no comprehensive 

study of neutron flux at LLNL; the neutron to photon ratio cannot be established using existing data 

from LLNL. A review of other DOE facilities that handled plutonium reported a neutron to photon ratio 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 14 of 24

Table 6-5. Interpretation of reported data. 

Period

Reported 

quantity Description

Interpretation 

of zeroes

Interpretation 

of blanks 

(no data)

Rollup of 

individual and 

annual data

Monitored/ 

unmonitored

1952–

1963

rem Reported WB 

doses include 

gamma and beta, 

sometimes 

qualified with “G” 

for gamma only. 

Neutron doses 

were designated 

with “N.”

Zeroes were 

generally not 

reported. Reported 

zero should be 

interpreted as 

meaning less than 

MDL.

Blanks should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB 

dose.

Neutron WB 

dose.

Shallow skin 

dose.

Total deep WB 

dose.

All employees with 

significant exposure 

potential were 

monitored for both 

gamma and neutron 

radiation (more than 

95% of employees 

were monitored 

before 1958). 

1963–

1985

rem Reported WB 

doses qualified as 

either “G” for 

gamma or “N” for 

neutron. Beta 

reported with “S” 

and/or “SK” for 

skin. 

Zeroes were 

generally not 

reported before 

1980. However, for 

any year, a blank or 

reported zero result 

should be interpreted 

as meaning less than 

MDL.

Blanks should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB 

dose.

Neutron WB 

dose.

Shallow skin 

dose.

Total deep WB 

dose.

All employees with 

significant measurable 

exposure potential

were monitored 

continuously.

1985–

present

rem Photon deep, 

neutron deep, and 

shallow dose 

reported. 

Zeroes were typically 

reported. Reported 

zero should be 

interpreted as 

meaning less than 

MDL.

Blanks should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB 

dose.

Neutron WB 

dose.

Shallow skin 

dose.

Total deep WB 

dose.

All employees with 

significant measurable 

exposure potential

were monitored 

continuously.

to range from 0.7 to 0.91. The study at Savannah River indicated the greatest ratio and a value of 1 is 

recommended for use for LLNL while the NTA film was used (ORAUT 2005c). The neutron to photon 

ratio applicable to LLNL was determined to have a geometric mean of 1.0, geometric standard 

deviation of 3.0, and an upper 95th percentile of 6.1. 

6.3 ADJUSTMENTS TO RECORDED DOSE

Photon Dose

No adjustment to recorded photon doses is recommended. Prior to 1986, dosimeters might have 

been calibrated in air (i.e., no phantom). Dose reconstructors should use roentgen-to-organ dose 

conversion factors (DCFs) when assessing reported photon doses prior to 1986. Since 1986, deep 

dose equivalents at LLNL have been based on DOELAP calibration to Hp(10). Use Hp(10)-to-organ 

DCFs beginning in 1986. Table 6-6 lists adjustments to recorded dose [5]. 

Neutron Dose 

Neutron exposure reported prior to 1969 was likely under-estimated. The NTA film was inadequate to 

retrospectively assess an estimate of the neutron dose that is assuredly favorable to claimants. The 

use of a neutron to photon ratio is described in Section 6.2.6 of this document.

To ensure favorability to the claimant, dose reconstructors can generally consider the neutron 

energies at LLNL to be between 0.1 and 2.0 MeV, for which the ICRP Publication 60 radiation 

weighting factor is 20 (ICRP 1991). The associated dose correction factor is 1.91. Dose 

reconstructors should apply this factor to measured neutron dose equivalent and missed neutron dose 

equivalent. See Table 6-6 [6]. 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 15 of 24

Table 6-6. Adjustments to recorded dose. 

Period Dosimeter Facility Adjustment to reported dose

1952–1985 Photon dosimeters All facilities Use roentgen-to-organ dose conversion factors. 

1986–present Photon dosimeters All facilities Use Hp(10)-to-organ dose conversion factors.

All years Neutron dosimeters All facilities Multiply reported doses by factor of 1.91 to account for 

ICRP 60 weighting factors (ICRP 1991; ORAUT 2006d).

6.4 MISSED AND UNMONITORED DOSE

The potential for missed dose exists when workers are exposed to radiation at levels below the 

detection limit of their personnel dosimeters. In the early years of radiation monitoring, when relatively 

high detection limits are combined with short monitoring durations, missed doses can be significant. 

This section discusses methodologies for estimating missed doses.

Unmonitored dose pertains to the potential dose received by workers who did not wear personnel 

dosimetry. LRL (ca. 1954) documented that all workers at Livermore were monitored for dose 

beginning in 1954. However, review of individual worker’s files indicates gaps in monitoring. This is 

evident for employees with low dose potential [7]. 

Watson et al. (1994) examined methods analysts can consider when there is no recorded dose for a 

period during a working career. The missed dose for dosimeter results less than the MDL is 

particularly important for earlier years when MDLs were higher and dosimeter exchange was more 

frequent. NIOSH (2006) describes options to calculate missed dose for this situation. The preferred 

option estimates a potential missed dose favorable to claimants as MDL/2 multiplied by the number of 

zero-dose results. Table 6-1 lists the results of these calculations. The analysis of co-worker data is 

the best estimate of unmonitored dose however, the analysis is not available at this time. A review of 

the co-worker study at the Los Alamos National Laboratory (LANL) indicated that exposure data 

exhibited a lognormal distribution and a geometric standard deviation (GSD) ranging from 2.7 to 5.82 

(ORAUT 2005d). The dose reconstructor should assume the LLNL results to have a lognormal 

distribution and a geometric standard deviation (GSD) of 4.0. The use of data from LANL provides an 

example of similar work performed with similar materials where the same type of controls were 

applied. The regulations for personnel exposure were the same for LLNL and LANL. In the absence 

of actual data from LLNL, the use of data from LANL is a reasonable attempt at providing estimates of 

dose distributions. The geometric mean and associated GSD for external exposures were provided in 

Table 6E-2 of the LANL TBD (ORAUT 2005d). The data were provided by each year of operation. It 

should only be used when no other LLNL specific data are available.

Specific incident reports might address significant nonroutine worker doses, such as skin 

contamination events. The dose assessments in such reports, based on investigations conducted at 

the time of the incident, should be the best estimates of dose received.

6.5 BIAS AND UNCERTAINTY

A number of factors contribute to uncertainty in measured doses. Systematic errors can occur from 

calibration and processing as well as from extraneous conditions such as moisture, heat, and fading. 

Random errors arise from variations among workers and the energy spectra and geometries of their 

exposures. This TBD analysis found no specific uncertainty assessments for LLNL dosimeter 

systems. However, the systems have much in common with systems used at other facilities, such as 

the DOE Hanford Site, for which extensive studies have been performed (Wilson et al. 1990; Fix, 

Gilbert, and Baumgartner 1994). The similarities enable reasonable comparisons for the LLNL 

systems based on experience at Hanford and elsewhere. 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 16 of 24

NIOSH (2006) provides guidance for estimating uncertainty in external dose reconstruction. Under 

good laboratory conditions, film-badge uncertainty can be at the level of 10% to 15%. The absolute 

uncertainty at 95% confidence should not be less than the MDL, which for LLNL, was as high as 

0.03 rem for beta/gamma film (Table 6-1). Figure 2.1 of NIOSH (2006) shows the results from two 

methods of calculating the uncertainty factor. In the absence of any other site-specific data, dose 

reconstructors should use numerical values employed for film-badge dose reconstruction [8]. 

An uncertainty factor of 1.2 is appropriate for dosimetry results based on the simplified dosimetry 

uncertainty modeling for film contained within NIOSH (2006). The uncertainty for TLDs is generally 

smaller than that for film and somewhat less dependent on energy. Dose reconstructors should use 

an uncertainty factor of 1.2, which is favorable to claimants for doses measured with both film and 

TLDs [9]. A reasonable option consistent with other TBDs and in dose reconstruction is to assume an 

uncertainty of ±30% for film and of ±20% for TLD. It is recommended that dose reconstructors 

assume a constant input to IREP by multiplying the measured dose by 1.3 for film and 1.2 for TLD.

Relatively little information is available on uncertainty for shallow dose. In view of the similar 

mechanisms between photon and beta film dosimetry, NIOSH (2006) recommends applying the 

methodology described above to the beta dose.

Table 6-7 lists bias and uncertainty factors for LLNL.

Table 6-7. Bias and uncertainty [10]. 

Site-specific dosimetry system

Bias magnitude and range

Overall biasa Range in biasb

Photon/electron film (1952–1969) 1.0 0.7–1.3

Neutron NTA film (1952–1969) c c

Harshaw photon/electron/neutron TLD (1969–1985) 1.0 0.8–1.2

Panasonic photon/electron TLD (1986–present) 1.0 0.8–1.2

CR-39 neutron (1986–present) 1.0 0.6–1.5

a. Based on the distribution of energy levels and geometry judged most likely. Divide recorded 

dose by the table's bias value to calculate deep dose.

b. Range of overall bias factors based on alternative distributions of energy levels and geometry.

c. Neutron dose for this time period should be based on the neutron to photon dose ratio. There 

is documentation that NTA film cannot provide reliable estimates of neutron dose.

6.6 DOSE RECONSTRUCTION

As much as possible, the basis for dose to individuals should be the dosimetry records. It is important 

to distinguish between the recorded non-penetrating and penetrating doses and the actual Hp(0.07)

and Hp(10). The following list contains appropriate guidance:

• Worker dosimetry records that provide nonzero beta-photon values for Hp(10) and Hp(0.07)

are adequate. Dose reconstructors should consider beta energies to be greater than 15 keV 

and, unless known to be otherwise, photon energies to be within the energy range of 30 to 250 

keV which is favorable to claimants [11]. 

• Radionuclides used in the various buildings and facilities are described in the LLNL site

description (ORAUT 2005e.) These will give the dose reconstructor additional evidence as to 

whether shallow doses are primarily from beta or low-energy X-rays if a particular building is 

cited in the workers' history files. If no evidence can be found as to the source of shallow 

dose, an approach that is favorable to claimants should be taken. In general, for most nonskin cancers, this results in assigning shallow dose to the 30- to 250-keV photons; for skin 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 17 of 24

cancers on exposed skin, assigning the dose to beta particles is more favorable to claimants

[12]. 

• Dose reconstructors should assign missed dose for workers for whom dosimetry records 

provide zero beta-photon values for Hp(10) and Hp(0.07) as identified in NIOSH (2006). This 

approach is conservative based on a review of historical data [13]. 

• For unmonitored workers, dose reconstructors should use the guidance in ORAUT-OTIB-0020 

(ORAUT 2005f) to apply a dose based on coworker data [14]. 

• For unmonitored workers whose exposure potential has been determined to be low, assign the 

environmental dose [15]. 

• The LLNL dosimetry programs were designed to measure the non-penetrating exposure to 

workers. The guidance in ORAUT-OTIB-0017 (ORAUT 2005a) should be used to reconstruct 

the non-penetrating dose.

• No evidence could be found to indicate whether elevated ambient levels of external radiation 

were subtracted from workers’ doses. However, because of the LLNL’s facility type, it is 

unlikely that there would have been a significant elevated ambient dose (ORAUT 2003b). 

6.7 ORGAN DOSE

NIOSH (2006) discusses the conversion of measured doses to organ dose equivalent; Appendix B of 

that document contains the appropriate DCFs for each organ, radiation type, and energy range based 

on the type of monitoring performed. In some cases, simplifying assumptions are appropriate. For 

periods when calibrations were performed in free air, prior to 1986, dose reconstructors should use 

the exposure-to-organ DCF. For recorded doses from 1986 to the present, use the Hp(10)-to-organ 

DCF.

6.8 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section with information that identifies the source and 

justification for each item. Conventional references are provided in the next section that link data, 

quotations, and other information to documents available for review on the Oak Ridge Associated 

Universities (ORAU) Team servers.

[1] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

Mr. Thomas reviewed the types of activities described in ORAUT (2005e) to establish the type 

of dosimeter, exchange frequency, and associated MDL and missed dose. The MDL was 

estimated by the dosimeter technology and the exchange period. When possible, the annual 

missed dose was calculated using the method selected by MDL/2 in OCAS-IG-001 (NIOSH 

2006). From 1952 to 1955, the dosimeters were issued and processed by LBNL.

[2] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006.

Mr. Thomas reviewed the types of activities described in ORAUT (2005e). Given the type of 

operations and the types of radioactive isotopes reported for each building, the energy 

fractions were selected in a manner similar to those for the LANL facilities.
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 18 of 24

[3] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

Mr. Thomas reviewed the types of activities that involved isotopes and tasks in which neutron 

radiation could be encountered as described in ORAUT (2005e). Given the type of operations 

and the types of radioactive isotopes reported for each building, the neutron energy selection 

of 0.1 – 2.0 MeV indicated in Table 6-3 was chosen because neutron sources were shielded 

and moderated to this energy range; the energy range of 0.1 – 2.0 MeV has the highest quality 

factor, as provided in 10 CFR 835.2, which is favorable to claimants. 

[4] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. Mr. 

Thomas reviewed the original data from multiple claims in the NOCTS database and made the 

observation that codes assigned to external dosimetry were updated over time. The types of 

dosimeters were described in the documentation provided with the exposure data.

[5] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

Before 1986, external dosimeters were calibrated in air without the use of a phantom. The 

“roentgen to organ” dose is adequately calculated using the DCFs used by the dose 

reconstructors. In 1986, LLNL changed its calibration procedures to use DOELAP 

requirements. The adjustments listed in Table 6-6 are consistent with other site technical 

basis documents, including those for LANL and RFETS. 

[6] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

Adjustments to neutron exposure are described in ICRP Publication 60 for neutron energies 

between 0.1 and 2.0 MeV (ICRP 1991). The dose correction factor of 1.91, reported in ICRP 

Publication 60, should be used to provide a correction to reported exposures and the 

calculation of missed dose. 

[7] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. March 2007. 

Review of NIOSH-Office of Compensation Analysis and Support Claims Tracking System

(NOCTS) cases, which included the Position/Titles of “Computer Administrator,” 

“Administrative Assistant,” “Administrative Specialist,” and “Secretary,” all showed 

unmonitored periods.

[8] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

The uncertainty factor for film badge dose reconstruction is provided by OCAS-IG-001 (NIOSH 

2006). This approach provides a series of consistent, assumptions for each DOE site, 

including LLNL, that are favorable to claimants. In the event that LLNL provided specific 

uncertainty factors for a particular film badge or batch of dosimeters, dose reconstructors

should use the NIOSH assumptions. 

[9] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

OCAS-IG-001 (NIOSH 2006) indicates that the uncertainty for TLD results is lower than the 

uncertainty assumed for film. The uncertainties of 1.3 for film and 1.2 for TLDs were derived 

from OCAS-IG-001 and should be assigned to film and TLD results at LLNL.

[10] Smith, Matthew H. ORAU Team. Health Physicist. May 2006. 

Mr. Smith reviewed the bias described for the types of dosimeters and LLNL facilities. He

recommended a revision of the table to be consistent with the Hanford TBD (ORAUT 2006c). 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 19 of 24

[11] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

Some dosimetry results do not distinguish between penetrating and nonpenetrating dose. The 

dose reconstructor is reminded to use an energy range dose that is favorable to claimants for 

nonpenetrating photons and a beta energy that will penetrate the outer layer of skin.

[12] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

The guidance provided describes two calculations by the dose reconstructor, exposed skin 

and covered skin. Some dosimetry results do not distinguish between penetrating and 

nonpenetrating dose. The dose reconstructor is reminded to assign a dose estimate dose 

which is favorable to claimants. 

[13] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

Some dosimetry results do not distinguish between penetrating and nonpenetrating dose. The 

assignment of missed dose that is described in OCAS-IG-001 is considered to be 

conservative. A review of historical data in NOCTS indicates that the missed dose assigned in 

OCAS-IG-001 (NIOSH 2006) is higher than the dose that would be assigned using the 

coworker data.

[14] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

The purpose of ORAUT-OTIB-0020 (ORAUT 2005f) is to provide general information to enable

ORAU Team dose reconstructors to assign doses to workers at DOE sites who have little or 

no individual monitoring data.

[15] Thomas, Bill R. Integrated Environmental Management. Health Physicist. April 2006. 

The radiation dose for unmonitored workers is characterized by ORAUT-OTIB-0020 (ORAUT 

2005f) and the external dose assigned by the environmental dose described in ORAUT 

(2005g). A low exposure potential is assumed to be a task in which an employee did not 

routinely handle radioactive materials and might have worked in an office environment and/or 

an administrative role.
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 20 of 24

REFERENCES

DOE (U. S. Department of Energy), 1986, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE Laboratory Accreditation Program for 

Personnel Dosimetry Systems, DOE/EH-0027, Assistant Secretary for Environment, Safety, 

and Health, Washington, D.C., December. [SRDB Ref ID: 12294]

Fix, J. J., E. S. Gilbert, and W. V. Baumgartner, 1994, An Assessment of Bias and Uncertainty in 

Recorded Dose from External Sources of Radiation for Workers at the Hanford Site, PNL10066, Battelle Memorial Institute, Pacific Northwest Laboratory, Richland, Washington, 

August. [SRDB Ref ID: 305]

Fix, J. J., L. Salmon, G. Cowper, and E. Cardis, 1997, "A Retrospective Evaluation of the Dosimetry 

Employed in an International Combined Epidemiologic Study," Radiation Protection Dosimetry, 

volume 74, pp. 39–53. [SRDB Ref ID: 7769]

Griffith, R. V., D. E. Hankins, R. B. Gammage, and L. Tommasino, 1979, “Recent Developments in 

Personnel Neutron Dosimeters - A Review,” Health Physics, volume 36, number 3, pp. 235–

260. [SRDB Ref ID: 7965]

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 1996, Conversion Coefficients for Use in 

Radiological Protection Against External Radiation, Publication 74, Pergamon Press, Oxford, 

England.

ICRU (International Commission on Radiation Units and Measurements), 1998, Conversion 

Coefficients for Use in Radiological Protection Against External Radiation, Report 57, 

Bethesda, Maryland.

LRL (Lawrence Radiation Laboratory), ca. 1954, Personnel Monitoring Procedure for all UCRL, 

Livermore, University of California, Livermore, California. [SRDB Ref ID: 23478]

Morgan, K. Z., 1961, “Dosimetry Requirements for Protection from Ionizing Radiation,” Selected 

Topics in Radiation Dosimetry, Proceedings of the Symposium on Selected Topics in 

Radiation Dosimetry, Sponsored by and Held in Vienna 7-11 June 1960, International Atomic 

Energy Agency, Vienna, Austria, pp. 3–23. [SRDB Ref ID: 8345]

NIOSH (National Institute for Occupational Safety and Health), 2006, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 2, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 25.

Nolan, W. E., 1958, “Progress Report on Film Badge Program,” memorandum to D. C. Sewell, 1958, 

University of California, Lawrence Radiation Laboratory, Livermore, California, November 13. 

[SRDB Ref ID: 23005]

ORAUT (Oak Ridge Associated Universities Team), 2003, Technical Information Bulletin: 

Occupational Dose from Elevated Ambient Levels of External Radiation, ORAUT-OTIB-0007, 

Rev. 00, Oak Ridge, Tennessee, November 12.
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ORAUT (Oak Ridge Associated Universities Team), 2005a, Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev. 01, Oak Ridge, Tennessee, October 

11.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Assignment of Missed Neutron Doses 

Based on Dosimeter Records, ORAUT-OTIB-0023, Rev. 00, Oak Ridge, Tennessee, March 7.

ORAUT (Oak Ridge Associated Universities Team), 2005c, Savannah River Site, ORAUT-TKBS0003, Rev. 03, Oak Ridge, Tennessee, April 5.

ORAUT (Oak Ridge Associated Universities Team), 2005d, Los Alamos National Laboratory –

Occupational Internal Dose, ORAUT-TKBS-0003, Rev. 02, Oak Ridge, Tennessee, May 10.

ORAUT (Oak Ridge Associated Universities Team), 2005e, Lawrence Livermore National Laboratory 

– Site Description, ORAUT-TKBS-0035-4, Rev. 00 PC-1, Oak Ridge, Tennessee, October 26.

ORAUT (Oak Ridge Associated Universities Team), 2005f, Use of Coworker Dosimetry Data for 

External Dose Assignment, ORAUT-OTIB-0020, Rev. 01, Oak Ridge, Tennessee, October 7.

ORAUT (Oak Ridge Associated Universities Team), 2005g, Lawrence Livermore National Laboratory 

– Occupational Environmental Dose, ORAUT-TKBS-0035-4, Rev. 00, Oak Ridge, Tennessee, 

November 18.

ORAUT (Oak Ridge Associated Universities Team), 2006a, Site Profile for the Lawrence Berkeley 

National Laboratory, ORAUT-TKBS-0049, Rev. 00, Oak Ridge, Tennessee, August 7.

ORAUT (Oak Ridge Associated Universities Team), 2006b, External Dose Reconstruction, ORAUTPROC-0006, Rev. 00 PC-3, Oak Ridge, Tennessee, December 11.

ORAUT (Oak Ridge Associated Universities Team), 2006c, Hanford Site – Occupational External 

Dosimeter, ORAUT-TKBS-0006-6, Rev. 02, Oak Ridge, Tennessee, November 21.

Pardue, L. A., N. Goldstein, and E. O. Wollan, 1944, Photographic Film As a Pocket Radiation 

Dosimeter, CH-1553-A-2223, University of Chicago, Metallurgical Laboratory, Chicago, Illinois, 

April. [SRDB Ref ID: 8599] 

Taylor, L. S., 1971, Radiation Protection Standards, CRC Press, Cleveland, Ohio.

Trost [illeg.], M., 2005, "LLNL Building Index: Pre 1966 and Post 1966," facsimile to J. J. Maisler 

(Integrated Environmental Management), University of California, Lawrence Livermore 

National Laboratory, Livermore, California, February 15. [SRDB Ref ID: 16948]

Watson, J. E., Jr., J. L. Wood, W. G. Tankersley, and C. M. West, 1994, "Estimation of Radiation 

Doses for Workers without Monitoring data for Retrospective Epidemiologic Studies," Health 

Physics, volume 67, number 4, pp. 402–405. [SRDB Ref ID: 8601]

Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

Hanford Personnel Dosimeter Program from 1944 through 1989, PNL-7447, Battelle Memorial 

Institute, Pacific Northwest Laboratory, Richland, Washington, September. [SRDB Ref ID: 

4793]
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 22 of 24

GLOSSARY

beta radiation 

Radiation consisting of electrons emitted spontaneously from the nuclei of certain radioactive 

elements. Most (if not all) direct fission products emit beta radiation. The beta particle is 

physically identical to an electron moving at high velocity.

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

deep dose equivalent (Hd) 

The dose equivalent at a depth of 10 millimeters in tissue.

dose equivalent (H) 

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The special unit is the rem. When D is expressed in grays, H is in sieverts (1 sievert equals 

100 rem). 

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See film dosimeter, neutron 

film dosimeter, thermoluminescent dosimeter.)

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation.

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, or 

extremities. This includes the fabrication, assignment, and processing of dosimeters as well 

as interpretation and documentation of the results.

film

In the context of this document, a packet that contains one or more pieces of film in a lighttight wrapping. The film, when developed, has an image caused by radiation that can be 

measured using an optical densitometer. (See nuclear emulsion.)

film dosimeter

A small packet of film within a holder that attaches to a wearer.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus.

minimum detection level (MDL)

The minimum quantifiable dose equivalent that a given dosimetry system can reliably 

measure.
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 23 of 24

missed dose

The potential dose equivalent that might not have been measured due to the limitation of the 

dosimeter even though a worker was monitored.

neutron

A basic particle that is electrically neutral and has nearly the same mass as the hydrogen 

atom.

neutron film dosimeter

A film dosimeter that contains a nuclear track emulsion, type A, film packet.

nuclear track emulsion, Type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused by neutrons 

that can be seen by using an appropriate imaging capability such as oil immersion and a 

1,000-power microscope or a projection capability.

occupational dose

The radiation dose resulting from a claimant’s exposure to radiation and/or to radioactive 

material while working at the LLNL site. Occupational dose does not include, for example,

exposure to background radiation (including radon), dose as a patient from medical practices

(including for treatment of work related injuries), dose from voluntary participation in medical 

research programs, or dose received as a member of the general public.

personal dose equivalent Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 millimeter and 

10 millimeters for the skin and body, respectively. These are noted as Hp(0.07) and Hp(10), 

respectively.

photon

A unit or particle of electromagnetic radiation consisting of X- and/or gamma rays.

rad

The traditional unit of absorbed dose (one rad equals 100 ergs per gram of material absorbing 

the radiation energy). The word derives from radiation absorbed dose. 

radiation

Alpha, beta, neutron, and photon radiation.

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

rem

The traditional unit of dose equivalent, which is equal to the product of the absorbed dose in 

rad and the quality factor of the radiation. The word derives from roentgen equivalent in man. 

roentgen

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X-) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kilogram of dry 
Document No. ORAUT-TKBS-0035-6 Revision No. 01 Effective Date: 04/26/2007 Page 24 of 24

air. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in 

soft tissue for higher (more than about 100 kilovolts-electron) energy photons.

shallow absorbed dose (Ds) 

The absorbed dose at a depth of 0.07 millimeter in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 millimeter in tissue.

sievert (Sv)

The International System unit for dose equivalent (1 sievert equals 100 rem). 

thermoluminescence

Property of a material that causes it to emit light as a result of being excited by heat.

thermoluminescent dosimeter (TLD)

A device used to measure radiation dose. It consists of a holder containing solid chips of 

material that when heated will release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose.

unmonitored dose

The potential unrecorded dose equivalent that could have resulted because an exposed 

worker was not monitored.

whole-body (WB) dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 centimeter (1,000 milligrams 

per square centimeter); however, this term is also used to refer to the recorded dose.

X-ray

(1) Ionizing electromagnetic radiation of external nuclear origin. (2) A radiograph.